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Amend 12 to License NPF-42,revising Tech Spec 3/4.3.1 in Area of Specified Surveillance Intervals & out-of-svc Times for Reactor Protection Sys Instrumentation
	ML20236M682
Person / Time
Site:	Wolf Creek
Issue date:	11/02/1987
From:	Calvo J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20236M684	List: ML20236M682; ML20236M706;
References
	NUDOCS 8711130243
	Download: ML20236M682 (18)
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UNITED STATES -

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_ KANSAS GAS & ELECTRIC COMPANY KANSAS CITY POWER AND LIGHT COMPANY

~ KANSAS ELECTRIC POWER COOPERATIVE, INC.

WOLF CREEK GENERATING STATION O,

DOCKET NO. 50-482 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.12 License No. NPF-42 i

1.

LThe Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Wolf Creek Generating Station (the. facility)' Facility Operating License No. NPF-42 filed by Kansas Gas and Electric Company acting for itself and Kansas City Power and Light Company and Kansas Electric Power Cooperative, Inc., (licensees) dated June 16, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commis-sion's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as

-amended, the provisions of the Act, and the rules and regulations of the Commission; t

C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety ~ of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and l

E.

The issuance of this amendment is in accordance with 10 CFF. Part 51 L

of the Commission's regulations and all applicable requirements have been satisfied.

l-8711130243 871102 PDR ADOCK 05000482 P

PDR l

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! -l 2.

Accordingly, the license is amended by changes to the Tech'nical Specifi-j cations as indicated in~the attachment to thisLlicense amendment and Paragraph 2.C.(2) of: Facility Operating License No. NPF-42 is hereby amended to read as follows:

2.

Technical Specification' u

The Technical Specificppions contained in Appendix A,.as revised through Amendment No. '.., and the Environmental Protection Plan contained in Appendix B; both of which are attached hereto, are hereby incorporated in the license.

KG&E shall operate the facility in accordance.with the Technical Specifications and the u

Environmental Protection' Plan.

~3.

'This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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$c4 M

/

i Jose A. Calvo, Director Project. Directorate - IV Division of Reactor Projects - III, IV, V and Special Projects-Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications-1 Date of Issuance: November 2,1987 I

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ATTACHMENT TO LICENSE AMENDMENT NO.12 FACILITY OPERATING LICENSE NO. NPF-42 DOCKET NO. 50-482 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE PAGES INSERT PAGES 3/4 3-3 3/4 3-3 3/4 3-5 3/4 3-5 3/4 3-6 3/4 3-6 3/4 3-9 3/4 3-9 3/4 3-10 3/4 3-10 3/4 3-11 3/4 3-11 3/4 3-12 3/4 3-12 8 3/4 3-1 B 3/4 3-1 B 3/4 3-2 through B 3/4 3-2 through B 3/4 3-6 B 3/4 3-7 (Reissued due to repagination, nochanges)

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TABLE 3.3-1 (Continued)

TABLE NOTATIONS

0nly if the Reactor Trip System breakers happen to be in the closed position and the Control Rod Drive System is capable of rod withdrawal.

- The borcn dilution flux doubling signal may be blocked during reactor startup in accordance with normal operating procedures.

The provisions of Specification 3.0.4 are not applicable.

1. Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

1. 1. Below the P-10 (Low setpoint Power Range Neutron Flux Interlock) Setpoint.

(1)The applicable MODES and ACTION statements for these channels noted in i

Table 3.3-3'are more restrictive and therefore applicable.

4 ACTION STATEMENTS ACTION 1"- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel j

to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within a

the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , j

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are' satisfied:

a.

The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; l

b.

The Minimum Char.nals UPERABLE requirement is met; however, j

the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months l

1 for surveillance testing of other channels per Specification 4.3.1.1; and c.

Either, lHERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron i

Flux-Trip Setpoint is reduced to less than or equal to j

85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the l

QUADRANT POWER TILT RATIO is monitored at least once per j

12 hcurs per Specification 4.2.4.2.

)

l ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a.

Below the P-6 (Intermediate Range Neutron Flux Interlock) 1 Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint; or b.

Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement suspend all operations involving positive reactivity changes.

WOLF CREEK - UNIT 1 3/4 3-5 Amendment No.12 1

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TABLE 3.3-1'(Continued)

ACTION STATEMENTS (Continued)-

~

. ACTION'S a.

With'the number of OPERABLE channels one less'than the Mini -

mum Channels.0PERABLE requirement,' restore the inoperable channel to OPERABLE status'within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the.Reac-tor Trip Breakers, suspend'all operations involving positive i

reactivity. changes and verify valves BG-V178 and BG-V601 are closed and secured in position within the next^ hour, b.

With'no channels OPERABLE, open the-Reactor Trip Brea.kers, suspend'all operations involving positive reactivity changes and verify compliance with the SHUTDOWN. MARGIN requirements.

of. Specification 3.1.1.1or 3.1.1.2, as applicable, within-1 hour and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, and verify valves BG-V178 and_BG-V601 are closed and secured in position within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and verified to be closed and secured in i

position every 14 days.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number'of Channels, STARTUP and/or POWER OPERATION may proceed provided the. following conditions are satisfied:

l a.

The inoperable channel is placed.'in the tripped condition.

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and l

q b.

The Minimum Channels OPERABLE.. requirement is met; however,

]

the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months j.

1 for surveillance testing of other channels per Specification 4.3.1.1.

I

-ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within i

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation'of the associated permissive annunciator window (s) that the. interlock is in its required state for the' existing plant' condition, or apply. Specification 3.0.3.

j ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1, provided-the other channel is OPERABLE.

ACTION 10 - With the number of OPERABLE channels one less than the Minimum I.

Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor trip breakers within the next hour.

4 ACTION 11 - With the number of 0PERABLE channels less than the Total Number l'

of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

c I

l WOLF CREEK - UNIT 1 3/4 3-6 Amendment No.12 L

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TABLE NOTATIONS

0nly if the Reactor Trip System breakers happen to be closed and the control rod drive system is capable of rod withdrawal.

1. Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

1. 1. Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1) If not performed in previous 31 days.

l (2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER.

Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%.

The provisions of Speci-fication 4.0.4 are not applicable for entry into MODE 2 or 1.

l (3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15%

of RATED THERMAL POWER.

Recalibrates if the absolute difference is greater than or equal to 3%.

The provisions of Specification 4.0.4 are not applic-able for entry into MODE 2 or 1.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

I (5) Detector plateau curves shall be obtained, evaluated and compared to manu-facturer's data.

For the Intermediate Range and Power Range Neutron Flux l

channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER.

The provi-sions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(9) Quarterly surveillance in MODES 3*, 4* and 5* shall also include verifica-tion that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. Quarterly surveillance shall include verification of the Boron l

Dilution Alarm Setpoint of less than or equal to an increase of twice the count rate within a 10-minute period.

(10) Setpoint verification is not required.

(11) At least once per 18 months and following maintenance or adjustment of the Reactor trip breakers, the TRIF ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the Undervoltage and Shunt trips.

(12) At least once per 18 months during shutdown, verify that on a simulated Boron Oilution Doubling test signal the normal CVCS discharge valves will close and the centrifugal charging pumps suction valves from the RWST will open within 30 seconds.

(13) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.

(14) Each c'hannel shall be tested at least every 92 days on a STA.iGERED TEST BASIS.

(15) The surveillance frequency and/or MODES specified for these channels in Table 4.3-2 are more restrictive and, therefore, applicable.

WOLF CREEK - UNIT 1 3/4 3-12 Amendment No.12 l

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3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engir.eered Safety Features Actuation System instrumentation and interlocks ensure that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is main-tained to permit a channel to be out-of-service for testing or maintenance and (4) sufficient system functional capability is available from diverse,

parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility

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design for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses.

The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit.

A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service times for the Reactor Protection Instrumentation System," supplements to that l

i report, and the NRC's Safety Evaluation dated February 21, 1985.

Surveillance intervals and out of service times were determined based on maintaining and an l

appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

ESF response times specified in Table 3.3-5 which include sequential opera-l i

tion of the RWST and VCT valves (Notes 3 and 4) are based on values assumed in

?.he non-LOCA safety analysis.

These analyses take credit for injection of i

borated water from the RWST.

Injection of borated water is assumed not to occur until the VCT charging pump suction vsives are closed following opening of the RWST charging pump suction valves.

When the sequential operation of the i

RWST and VCT valves is not included in the response times (Note 7), the values specified are based on tt.e LOCA analyses.

The LOCA analyses take credit for i

injection flow regardless of the source.

Verification of the response times specified in Table 3.3-5 will assure that the assumptions used for the LOCA and non-LOCA analyses with respect to operation of the VCT and RWST valves are valid.

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To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured und calibrated, WOLF CREEK - UNIT 1 8 3/4 3-1 Amendment No. 9,12 J

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i REACTOR' TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) l I

. Allowable Values for the Setpoints have been specified in Table'3.3-4.

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with Setpoints less conservative than the Trip Setpoint but within the' Allowable-Value is acceptable since an allowance.has been made in the safety analysis to

. accommodate this error.

An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the fAllowableValue.

The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in

. calibrating the instrumentation.

In Equation 3.3-1, Z + R + S 5 TA, the

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' interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered.

Z, as specified in Table 3.3-4,

' in percent span, is the statistical summation of errors. assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement.

TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for the actuation.

'R or Rack Error is the."as measured" deviation, in percent span, for the

'affected-channel from the specified Trip Setpoint.

S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions.

The methodology to derive the Trip Setpoints is based upon combining all

'of the uncertainties in the channels.

Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties.

Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.

Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.

Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected.'

Rack or sensor drif t, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

The measurement of response time at the specified frequencies providea assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the safety analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demon-strated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either: (1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

The Engineered Safety Features Actuation System senses selected plant para-meters and determines whether or not predetermined limits are being exceeded.

If they are, the signals are combined into logic matrices sensitive to combina-tions indicative of various accidents, events, and transients.

Once the required logic combination is completed, the system sends actuation signals to those WOLF CREEK - UNIT 1 B 3/4 3-2

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i INSTRUMENTATION BASES l

l REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM l

INSTRUMENTATION (Continued)

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Engineered Safety Features components whose aggregate function best serves the requirements of the condition.

As an example, the following actions may be i

initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: (1) Safety Injection pumps start and automatic valves position, (2) Reactor trip, (3) Feed-water System isolates, (4) the emergency diesel generators start, (5) contain-ment spray pumps start and automatic valves position,-(6) containment isolates, (7) steam line isclation, (8) Turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment cooling fans start and automatic valves position, (11) essential service water pumps start and automatic valves position, and (12) isolate normal control room ventilation and start Emergency Ventilation System.

Engineered Safety Features Actuation System Interlocks The Engineered Safety Features Actuation System interlocks perform the following functions:

P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T below Setpoint, prevents the opening of the main avg feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level signal, allows safety Injection block so th?.t components can be reset or tripped.

Reactor not tripped prevents manual block of Safety injection.

P-11 On increasing pressure P-11 automatically reinstates safety injection actuation on low pressurizer pressure and low steamline pressure and automatically blocks steamline isolation on negative steamline pressure rate.

On decreasing pressure; P-11 allows the manual block of Safety Injection on low pressurizer pressure and low steamline pressure and allows steamline isolation on negative steamline pressure rate to become active upon manual block of low steamline pressure SI.

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3/4.3.3 MONITORING' INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT ~0PERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant-1 operations ensures that: (1) the associated ACTION will be initiated when the

< radiation level monitored by each channel or combination thereof reaches 'its Setpoint,- (2) the specified coincidence logic is maintained, and (3) suffi-

cient: redundancy is maintained to permit.a channel to be out-of-service 'for l

a testing or maintenance.. The radiation monitors for plant operations' senses r

radiation levels in selected plant systems.and locations and' determines.whether or not predetermined limits'are being exceeded..If they are, the signals are combined into logic matrices sensitive to combinations indicative of various

. accidents and abnormal conditions.

Once the required logic combination is completed,_the system sends actuation signals to initiate alarms or. automatic isolation action and acutation of Emergency Exhaust or Control Room Emergency Ventilation Systems.

.3/4.3.3.2 MOVABLE INCORE DETECTORS j

The OPERABILITY of the movable incore detectors with the specified minimum scomplement of. equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core.

The OPERABILITY of this system is demonstrated by irradiating each.

t detector.used and determining the acceptability.of its voltage curve.

For the purpose of measuring F (Z) or F a fuH iNore flux map is used.

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Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of'the Excore Neutron Flux Detection System, and full incore

-flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT l

POWER TILT RATIO when one Power Range Neutron Flux channel is inoperable.

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3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

This capability i

is required to permit comparison of the measured response to that used in the design basis for the facility to datermine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100.

The instrumentation is consistent i

with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquakes," April 1974.

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INSTRUMENTATION-34

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BASES 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The'0PERABILITY of the meteorological instrumentation ensures that

-sufficient meteorological data is available for estimating potential ~ radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide.l.23, "Onsite Meteorological Programs," February 1972.

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'3/4.3.3.5 REMOTE SHUTOOWN INSTRUMENTATION i

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The OPERABILITY of the Remote Shutdown System ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTOOWN of the facility from locations outside of the control room and that a fire will not preclude achieving safe shutdown.

The Remote Shutdown System transfer switches, power circuits, and control circuits are independent of areas where J

a fire could damage systems normally used to shut down-the reactor.

This capability is required in the event control room habitability is lost and is consistent with General Design Criteria.3 and 19 and Appendix R of 10 CFR Part 50.

3/4.3.3.6~ ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.

This capability is consistent with the recommendations of Regulatory Guide 1.97, Revision 2, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and

-t Following an Accident," December 1980 and NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.

3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the Chlorine Detection System ensures that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release.

This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," Revision 1, January 1977.

3/4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that both ade-quate warning capability is available for the prompt detection of fires and that Fire Suppression Systems, that are actuated by fire detectors, will discharge extinguishing agents in a timely manner.

Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an' integral element in the overall facility fire protection program.

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.4 INSTRUMENTATION BASES y

FIRE DETECTION INSTRUMENTATION (Continued)

Fire detectors that are used to actuate fire suppression systems represent a more critically important component of a plant's fire protection program than detectors that are installed solely for early fire warning and notification.

Consequently, the minimum number of operable fire detectors must be greater.

The loss of detection capability for Fire Suppression Systems, actuated by fire detectors represents a significant degradation of fire protection for any area.

As a result, the establishment of a fire watch patrol must be initiated at an earlier stage than would be warranted for the loss of detectors that provide only early fire warning.

The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

3/4.3.3.9 LOOSE-PART DETECTION INSTRUMENTATION The'0PERABILITY of the loose part detection instrumentation ensures-that sufficient capability is available to detect loose metallic parts in the Reactor Coolant System and avoid or mitigate damage to Reactor Coolant System components.

The allowable out-of-service times and Surveillance Requirements are consistent with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981.

3/4.3.3.10 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents.

The Alarm /

Trip Setpoints for these instruments shall be calculated and adusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

3/4.3.3.11 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous ef fluents.

The Alarm / Trip Setpoints for these instruments shall be adjusted to values calculated in accordance with the methodology and parameters in the 00CM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

The instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM.

The OPERABILITY and use of this instrumentation is consistent with the 6 pCi/cc are measurable.

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l RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION (Continued) requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR l

.Part 50 'The sensitivity of any noble gas activity monitor used to show compli-

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ance with the gaseous effluent release requirements of Specification 3.11.2.2

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shall be such that concentrations as low as 1 x 10-l 3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed pro-1 tection instrumentation and the turbine speed control valves are OPERABLE and

'will prntect the. turbine from excessive overspeed.

Although the orientation of the turbine-is such that the. number of potentially damaging missiles which could impact and damage safety-related components, equipment, or. structures is minimal, protection from excessive turbine overspeed is required.

J WOLF CREEK - UNIT 1 8 3/4 3-7

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ML20236M682

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